Disposal of Intermediate Level Waste - GNSSN Home · 2016-12-01 · No. ID Presenter Title of Paper...

Session 3c – ILW IAEA-CN-242

International Conference on the Safety of Radioactive waste Management

SESSION 3c

Disposal of

Intermediate Level Waste


2

ORAL PRESENTATIONS

No. ID Presenter Title of Paper Page

03c – 01 64 R. Nakabayashi

Japan

Development of Methodology for Probabilistic

Safety Assessment of Long Term Radioactive

Waste Disposal

4

03c – 02 91 S. Konopaskova

Czech Republic

Waste Acceptance Criteria Development for

Different Low and Intermediate Level Waste

(LILW) Disposal Systems

8

03c – 03 92 E. Andersson

Sweden

Assessment of Human Intrusion and Future

Human Actions – Example from the Swedish

Low and Intermediate Level Waste Repository

SFR

13

03c – 04 97 A. Carter

United Kingdom

Data Management to Support a Post-Closure

Safety Case for Higher Activity Wastes

17

03c – 05 193 H. Arlt

United States of

America

Greater-Than-Class C Low Level Radioactive

Waste Characteristics and Disposal Aspects

21

03c – 06 135 J.-M. Hoorelbeke

France

Implementation of a Graded Approach in

Radioactive Waste Management in France

26


3

POSTER PRESENTATIONS

No. ID Presenter Title of Paper Page

03c – 07 80 K. Källström

Sweden

Methodology and Results for the Safety

Assessment for Low and Intermediate level

Waste Repository (SFR) in Sweden

30

03c – 08 95 A. Glindkamp

Germany

Implementation of Requirements on the

Chemical Toxicity of Nuclear Waste at a

Repository

35

03c – 09 102 M. Nepeypivo

(A. Talitskaya)

Russian Federation

Safety Assessment as an Instrument for Waste

Acceptance Criteria Derivation

38

03c – 10 128 B. Samwer

Germany

Konrad Repository – Evaluation on the Safety

Requirements according to the State of the Art

of Science and Technology

43


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03c – 01 / ID 64. Disposal of Intermediate Level Waste

DEVELOPMENT OF METHODOLOGY FOR PROBABILISTIC SAFETY

ASSESSMENT OF LONG-TERM RADIOACTIVE WASTE DISPOSAL

R. Nakabayashi, D. Sugiyama

Central Research Institute of Electric Power Industry (CRIEPI), Tokyo, Japan

E-mail contact of main author: [email protected]

Abstract. This paper discusses a methodology for probabilistic safety assessment for long-term radioactive

waste disposal considering both the epistemic and aleatory uncertainties included in the safety assessment. This

methodology can be used to demonstrate compliance with dose criteria and be helpful for the optimization of

radiation protection in a waste disposal programme. In addition, the applicability of the probabilistic approach is

demonstrated by illustrating a safety assessment for a radioactive waste disposal facility in Japan.

Key Words: Probabilistic safety assessment; uncertainty; dose criteria; optimization

1. Introduction

For the protection of people after the closure of a disposal facility, the disposal facility has to

be designed so as not to exceed the dose constraint that is used as a dose criterion, and

radiation protection is required to be optimized [1]. In disposal of long-lived radioactive

waste, safety assessment must take into consideration not only aleatory uncertainties but also

epistemic uncertainties. An aleatory uncertainty originates from the inherent heterogeneity or

diversity of data (e.g., the fracture permeability of host rock), and an epistemic uncertainty is

due to lack of knowledge (e.g., the degradation time of an engineered barrier). In this paper,

we briefly review a methodology for probabilistic safety assessment considering both

epistemic and aleatory uncertainties [2]. This method was developed to determine compliance

with the dose criterion of 0.3 mSv/year and to provide useful material for the optimization of

radiation protection. In addition, the applicability of the probabilistic safety assessment is

demonstrated by illustrating a safety assessment for a radioactive waste disposal facility in

Japan.

2. Framework of the methodology of probabilistic safety assessment

2.1.Probabilistic dose assessment

The procedure of the probabilistic dose assessment [2] is briefly described as follows.

(1) Aleatory and epistemic uncertainties are quantified as a probability distribution by

applying a statistical process to measured data and eliciting expert judgments,

respectively.

(2) Probabilistic dose assessment is carried out in consideration of both the epistemic and

aleatory uncertainties by using the radionuclide migration program with a Monte Carlo

simulation.

(3) The cumulative distribution function (CDF) of the maximum annual dose of a certain

radionuclide is calculated in the dose assessment. The probability density function (PDF)

is also calculated by kernel density estimation.

mailto:[email protected]


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2.2.Demonstration of compliance with a dose criterion

It is possible to demonstrate the compliance with a dose criterion by comparing the 95th

percentile of the CDF with 0.3 mSv/year and analogically adopting the concept of the

‘representative person’ [3]. The ICRP recommends that the representative person should be

defined such that the probability is less than about 5% that a person drawn at random from

the population will receive a greater dose in a probabilistic safety assessment. This indicates

that the vast majority of the population is protected from the radiation when the 95th

percentile of the dose distribution incorporating the uncertainties involved is less than the

dose criterion. By demonstrating that the 95th

percentile of the dose distribution, obtained by

probabilistic dose assessment in consideration of the uncertainties associated with long-term

radioactive waste disposal, is less than 0.3 mSv/year, the aim of protection of the public is

achieved. In Figs. 1(a) and 1(b), the maximum annual dose C is adopted as the assessment

result for comparison with the dose criterion of 0.3 mSv/year.

2.3.Optimization of radiation protection

In the optimization of radiation protection through compliance with the dose criterion of 0.3

mSv/year, not only the 95th

percentile of the CDF but also the mode of the PDF should be

reduced to as low as reasonably achievable while taking economic and social factors into

account. The mode of the PDF is the most likely dose that the public will be exposed to,

which is derived from the most likely behavior of the disposal system. Efforts to reduce the

most likely dose lead to the increased safety of a waste disposal facility. If more than one

option is capable of providing the required level of safety (i.e., the 95th

percentile of the CDF

is less than 0.3 mSv/year), then other factors, which are economic and social, also have to be

considered [1]. We propose that the mode of the PDF is one of the most important factors in

addition to the 95th

percentile of the CDF for the optimization of radiation protection.

If a regulatory body sets out a dose criterion for the most likely behavior of a disposal system,

the determination of the compliance with the dose criterion should performed conservatively.

In this case, the larger of the modal value of the PDF and the 50th

percentile of the CDF can

be employed to meet the dose criterion as discussed in a previous paper [2]. In Fig. 1(a),

where dose B (mode of PDF) is greater than dose A (50th

percentile of CDF), dose B is

adopted as the assessment result for comparison with the dose criterion. In Fig. 1(b), where

dose B (mode of PDF) is less than dose A (50th

percentile of CDF), dose A is adopted as the

assessment result for comparison with the dose criterion.

FIG. 1. Concept of the approach to determine the compliance with dose criteria [2].


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3. Application of probabilistic safety assessment

This section outlines how we carry out a safety assessment for radioactive waste disposal in

Japan using the probabilistic approach. Note that this is only an example used to discuss the

applicability of the probabilistic approach.

The regulatory body in Japan requires applicants to demonstrate that their dose assessment

results are less than the dose criteria assigned for each scenario in consideration of the

uncertainties. The dose criteria of likely and less-likely scenarios, which are classified in

terms of their likelihood of occurrence based on a disaggregated dose/probability approach,

are 0.01 and 0.3 mSv/year, respectively [4]. The purpose of conducting the safety assessment

of the likely scenario is to evaluate whether the basic design of the disposal system has been

considered to minimize the effects of radiation on the public (i.e., less than 0.01 mSv/year)

under normally expected scenarios. The purpose of the less-likely scenario is to check

whether the doses based on the scenario are below the dose criterion of 0.3 mSv/year, even

when taking into account uncertainties that are less likely but may have a significant effect in

the safety assessment. This example presents the assessment of compliance with the likely

and less-likely scenarios in consideration of the epistemic and aleatory uncertainties

associated with a sub-surface disposal system.

We consider the dose assessment model for exposure pathways in groundwater migration in a

sub-surface disposal system and deal with the epistemic uncertainty concerning the

degradation times of the engineered barriers and the aleatory uncertainty concerning the

permeability coefficient of the host rock. The engineered barriers are composed of

cementitious or bentonite materials. 14

C (4.38×1015

Bq) is instantaneously released from

radioactive waste in this model.

3.1. Quantification of epistemic uncertainty and aleatory uncertainty

The epistemic uncertainty concerning the degradation time of each barrier was expressed as a

subjective probability distributions on the basis of expert judgment (Fig. 2). The aleatory

uncertainty concerning the permeability coefficient of the host rock was expressed as a log-

normal distribution by applying a statistical process to measured values (Fig. 2). For details

of the quantification, refer to Nakabayashi and Sugiyama (2016) [2].

FIG. 2. Probability distributions for the epistemic uncertainty concerning the degradation time of a

cementitious barrier (a) and bentonite barrier (b), and aleatory uncertainty concerning the

permeability coefficient of the host rock (c) used in the safety assessment [2].


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FIG. 3. PDF and CDF indicating the maximum annual dose of 14

C [2].

3.2. Determination of compliance with the likely and less-likely scenarios

The PDF and CDF of the maximum annual dose of 14

C were obtained from the probabilistic

safety assessment (Fig. 3). In this section, we illustrate how to demonstrate the assessment

results in compliance with the stepwise dose criteria (0.01 and 0.3 mSv/year) of likely and

less-likely scenarios. The mode of the PDF was 0.0025 mSv/year, whereas the 50th

percentile

of the CDF was 0.0029 mSv/year, i.e., the 50th

percentile was larger than the mode. In this

case, the 50th

percentile is adopted as the assessment result for comparison with the dose

criteria in the likely scenario. The 95th

percentile of the CDF is adopted as the assessment

result for comparison with the dose criteria of 0.3 mSv/year in the less-likely scenario.

4. Conclusion

A probabilistic safety assessment considering epistemic and aleatory uncertainties has been

proposed to determine the compliance with a dose constraint of 0.3 mSv/year. This

methodology can estimate the mode of the PDF, which is the most likely dose that the public

will be exposed to. For the optimization of radiation protection, it is important to strive to

reduce the 95th

percentile of the CDF and the mode of the PDF to as low as reasonably

achievable while taking economic and social factors into account.

REFERENCES

[1] International Atomic Energy Agency, Disposal of Radioactive Waste, IAEA Safety

Standards Series No. SSR-5, IAEA, Vienna (2011).

[2] Nakabayashi, R., Sugiyama, D., Development of Methodology of Probabilistic Safety

Assessment for Radioactive Waste Disposal in Consideration of Epistemic Uncertainty

and Aleatory Uncertainty, Journal of Nuclear Science and Technology, Taylor & Francis

(2016).

[3] International Commission on Radiological Protection, Assessing Dose of the

Representative Person for the Purpose of Radiation Protection of the Public and the

Optimisation of Radiological Protection, Publication 101, Pergamon Press, Oxford and

New York (2006).

[4] Nuclear Safety Commission of Japan, Basic Policy for Safety Regulations Concerning

Land Disposal of Low-Level Waste (Interim Report), NSC, Tokyo (2007).


8


WASTE ACCEPTANCE CRITERIA DEVELOPMENT FOR DIFFERENT LILW

DISPOSAL SYSTEMS

S. Konopaskova, D. Lukin, I. Zadakova

Radioactive waste repository authority (SURAO), Praha, Czech Republic


Abstract. In the Czech Republic, there are operated two types of repositories: near surface for disposal of low

level waste from NPPs, and repositories for institutional waste, specified as low and intermediate level waste;

these are located underground, in former mines of different types. New legislation after 1997 and optimized

conditions for final waste form characterization lead to improvement of WAC derivation methods by the means

of safety assessment and supported their variety.

Key Words: waste acceptance criteria, subsurface repository, repository for low and

intermediate level waste, safety assessment

1. Introduction

This paper describes the procedure of waste acceptance criteria (WAC) development, applied

for various types of operated radioactive waste repositories in the Czech Republic. Safety

related criteria are derived using the results of safety assessment, considering waste streams,

barriers system, and position of the repository in the host structure. Special considerations are

included evaluating hydrogeological conditions of the host structure and accessible

biosphere. Differences of repositories lead to differences in WAC, as it is presented below.

2. WAC for disposal systems in the Czech Republic

In the Czech Republic, there are operated two types of radioactive waste repositories:

Subsurface disposal of waste from nuclear power plants

Disposal of low and intermediate level institutional waste in mine cavities, some

tenths of meters below surface

WAC defined for individual repositories differ in extent, qualitative expression and

quantitative parameters thanks to specific approach to their derivation, considering different

project, operational and environmental conditions of the repositories.

2.1. Operated repositories and their types

The overview of repositories is specified in Table 1.



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TABLE I: MARGINS FOR YOUR MANUSCRIPT.

Site Type Volume Waste streams Matrix Host rock

Dukovany,

1993 -

subsurface 55 000 m3

Waste from NPPs Bitumen,

geopolymer

Crystalline

Richard,

1964 -

LILW 17 000 m3

Institutional waste,

artificial radionuclides

Cement Limestone

mine

Bratrství,

1973 -

LILW 1 200 m3 Institutional waste,

natural radionuclides

Cement Uranium mine

FIG. 1. Dukovany repository – a subsurface vault system

FIG. 2. Richard and Bratrství repositoris – underground disposal systems

2.2.WAC structure

WAC are structured according to safety requirements, technical restrictions and

administrative requirements defined by law.

Safety related criteria guarantee the compliance with qualitative and quantitative objectives

of nuclear safety and radiation protection. These criteria are derived from the results of safety

assessment, namely:

Total activity of radionuclides in the repository

Volume activity of radionuclides in different waste forms


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Leachability of the final waste form

Activity of radionuclides in non-solidified waste

Stress resistance of final waste form solidified by cement and/or geopolymer

Dose rate on the surface of waste package

Technical restrictions are done by repository construction and include:

Water presence in drainage system

Weight of waste package

Integrity and structural stability of the waste package

Administrative and formal restrictions guarantee the compliance with nuclear and

environmental legislation:

Presence of free liquids, pyrophoric and toxic substances, complexing and

microbiological agents

Waste tracking system and passport

2.3. Derivation of safety related WAC

In the procedure of safety related WAC, there are more aspects that can lead to differences in

WAC specifications, done by:

Composition of the radionuclide vector

Final waste form and packages properties

Repository construction

Depth of the repository below surface

Hydrogeological conditions of the host structure, and

Probable use of the land in communication points

Generally, there is defined a set of scenarios supporting safety case, i. e. operational scenarios

and long term scenarios that should help to evaluate probable radiation effects during

operations and after repository closure.

Normal evolution scenario is used to define the capacity of the site - volume of the waste as

well as its total activity. Scenario components are site specific. For subsurface system, direct

infiltration of rainwater and advective flow through disposal units are considered

immediately after institutional control period. For underground system, the infiltration is

controlled by diffusion and by inflow to fractures in near field and advective flow starts much

later thanks to final waste form and filling stability. Safety function of host structure is

strongly affected by hydrogeology system as a part of transport pathway in all types of

repositories.

Alternative scenarios are used to evaluate disposal system performance by deviations from

projected performance. For near surface repository, bathtubbing is considered; in mine

systems, possible outflow of contaminated mine water is taken into account.

Intrusion scenarios are the base for limiting volume activities in the final waste form. In

subsurface system, evaluation of on site residence and working activities on the site after

institutional control are considered. For underground repositories, there is evaluated a contact

with waste as a consequence of drilling activities.

Limits of dose rates in the controlled area and on the waste packages are derived by means of

radiation protection. In addition, radon intake has to be considered in underground

repositories.


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Potential emergency situations could lead to non acceptable doses of workers, but the

evaluation of accidents is not considered in formulating criteria for radioactive content of the

waste.

2.4. Quantitative comparison of safety related criteria

The limits of total and volume activities have been compared for various disposal systems, as

it is indicated in FIG. 1. In spite of the fact that the composition of radioactive waste and its

activities are not identical in waste from NPPs and in institutional waste, there are some

radionuclides present in both types of waste. The results of safety cases lead to lower

permitted activities and activity concentration in the vault system and the capacity.

Underground repositories provide higher capacity for activity of short lived radionuclides as

well as for long lived radionuclides, thanks to sophisticated stabilization system, better

hydrogeological conditions, as well as to lower potential for inadvertent intrusion.

FIG. 3 .Limits of activity in vault and underground systems

3. Conclusions

Normal evolution scenario is limiting for both total and volume activities in mine

repositories, Higher capacity of mine system is shown thanks to better engineered barriers

performance and/or lower probability of intrusion.

EBS system is more efficient in mine systems – barrier can assure diffusion driven transport

for longer periods of time even in the case that the number of fractures in near field is

relatively high.

In subsurface system, increasing of life time of barriers leads to higher doses in intrusion and

on site scenarios as a consequence of negligible decrease of activity of long lived

radionuclides during institutional control period. Volume activities in subsurface system have

to be strongly limited, to the values of very low level waste.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Derivation of activity limits for

the disposal of radioactive waste in near surface disposal facilities, IAEA-TECDOC-

1380, Vienna (2003).

[2] KONOPASKOVA, S., et al., “Safety report of Dukovany repository”, SÚRAO, Praha

(2012).


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[3] KONOPASKOVA, S., et al., “Safety report of Richard repository”, SÚRAO, Praha

(2014).

[4] MILICKY, M. et al., Hydrogeological model of the Dukovany site, ProGeo 2012

[5] MILICKY, M. et al., Hydrogeological model of the Richard site, ProGeo 2013

[6] Regulation SONS No. 307/2007 Coll. on radiation protection


13


ASSESSMENT OF HUMAN INTRUSION AND FUTURE HUMAN ACTIONS -

EXAMPLE FROM THE SWEDISH LOW AND INTERMEDIATE LEVEL WASTE

REPOSITORY SFR

E. Andersson1, T. Hjerpe

2, G. Smith

3, K. Källström

1, L. Morén

1, K. Skagius

1

1Swedish Nuclear Fuel and Waste Management Co., Sweden

2Facilia AB, Sweden,

3GMS Abingdon Ltd, UK


Abstract. The strategy commonly adopted in the disposal of solid radioactive waste is to contain the waste so

that it is kept away from the accessible biosphere by means of underground disposal. The intention is to isolate

the waste from man and the biosphere for a sufficiently long time to allow radioactive decay to significantly

reduce the radiation hazard. However, the potential exposure to radioactive material following intrusion is an

inescapable consequence of the deposition of the radioactive waste in a repository. There is an international

consensus that future human actions (FHA) resulting in disruption of the disposal facility must be considered in

the safety assessment as part of the safety case for a radioactive waste repository. However, although there are

some general recommendations concerning assessment of radioactive waste disposal, there is no over-arching

international methodological guide on how to perform FHA assessments. There is an ongoing project at IAEA

on handling inadvertent human intrusion (HIDRA). The Swedish Nuclear Fuel and Waste Management

Company (SKB) is taking part in the HIDRA project for human intrusion but also analyse the broader concept

FHA.

SKB has performed several analyses of FHA (including human intrusion by drilling) for both the existing

repository for low- and intermediate level waste (SFR) situated at 60-120 m depth and for the planned repository

for spent nuclear fuel to be situated at approximately 500 m depth. The SKB methodology to assess FHA

includes FEP-analysis, identification of stylised scenarios and qualitative and quantitative evaluation of the

stylised scenarios.

In December 2014, SKB submitted an application to the Swedish Radiation Protection Authority (SSM) to

extend the existing repository for low- and intermediate level waste (SFR). The planned extension includes 6

additional rock caverns to be placed at 120 m depth. The safety case for the application included an assessment

of FHA for both the existing part of the repository and for the planned extension. The methodology used and

major results of the FHA analysis are presented. In addition, examples are given of adjustments to general

recommendations that were needed to address FHA issues relevant to the assessment needs for this specific

assessment.

Key Words: Human intrusion, future human actions, waste disposal, safety assessment

1. Introduction

There is a long-standing international consensus that future human actions (FHA) and human

intrusion (HI) resulting in some disruption to the repository must be considered in safety

assessments as part of a safety case for a radioactive waste repository [1, 2, 3]. However,

there is no over-arching international guide on how to incorporate FHA in assessments. IAEA

has an ongoing project, ‘Human Intrusion in the context of Disposal of RAdioactive waste’

(HIDRA) to develop and test a methodology [4]. There are also other international projects

where experiences of handling HI in different countries have been shared e.g. [5] and further

commentary provided [6]. Depending on site specific and repository specific conditions as

well as regulatory and local stakeholder considerations, different aspects of FHA may need to



14

be considered. SKB has addressed FHA in safety assessment since the late 90’s. In this paper,

the methodology [7] used in the assessment of FHA for the low and intermediate level waste

(L/ILW) repository SFR is described.

2. General recommendations by international projects

Although an international FHA methodology is not available, there are useful

recommendations in documents like those mentioned above. Below is a list of some typical

examples and a note when they have not been followed in the SFR assessment [7].

Select a site away from natural resources in order to minimize likelihood for intrusion.

Only consider inadvertent intrusion, i.e. actions carried out when the location of the

repository is unknown, its purpose forgotten or the consequences of the actions are

unknown. Current society cannot be required to protect future societies from their

own intentional and planned activities if they are aware of the consequences.

A common approach to societal conditions is to use current conditions, both regarding

human behavior and technological development. Sites my change due to e.g. climate

change, then current data from sites with similar conditions may be used in the

assessment. In the area where SFR is situated, land uplift leads to areas currently

covered by sea to be situated below dry land. This has been addressed in the FHA

analysis.

Avoid quantitative use of probabilities because it is difficult to justify assigning a

number to the probability of specific FHA. Nevertheless, some quantitative

consideration of probabilities of such events is considered in the FHA assessment for

SFR.

Instead of trying to identify every possible feature, event and process (FEP) and

analyze all possible FHA, it is recommended to use a few stylized scenarios to

illustrate the range of consequences if they were to occur. However, a FEP-list is a

good tool to identify a consolidated set of relevant scenarios and this approach has

been used in the SFR assessment.

3. Relevant features of the repository SFR in Sweden

SFR is an existing repository for L/ILW situated below the sea floor in the Baltic Sea. The

sea is currently a barrier for HI but due to the ongoing post-glacial land uplift SFR will be

situated below land in the future and then HI will be possible. The assessment needed to

consider these altered future conditions at the site even though the geosphere remains an

effective barrier. SFR consist of 4 rock vaults and one silo situated between 60-120 m depth

in granitoid rock. In 2014, SKB applied to extend the repository with 6 rock vaults and filed a

safety assessment including assessment of FHA [7].

4. Methodology with examples from the SFR assessment

In the assessment of FHA for SFR, a step-wise methodology was used (Fig. 1).

5. Analysis of FEPs

A FEP-list was produced by first identifying safety relevant factors and then identifying

actions (FEPs) related to FHA that could negatively affect these safety factors. The audited

FEP-list proved to be a good tool for generating stylized scenarios and in communication


15

with the public. Our experience suggests that treating future human actions that the public are

concerned could pose a hazard to future generations in similar manner to the FEPs in the

main risk assessment can help to build confidence in the safety case.

FIG. 1. Overview of the stepwise methodology used for handling FHA at SKB.

6. Scenarios and calculation cases

Based on the FEP-list, a consolidated set of stylized scenarios was identified, taking into

account stakeholder interests. In this consolidation all FEPs were covered by at least one

scenario unless there were effective and documented arguments that the FEP would not affect

the robustness of the safety case. The scenarios evaluated were:

Drilling scenario including four separate calculation cases

o Exposure due to utilizing the drilling hole as a well

o Exposure to on-site crew during the drilling

o Exposure during construction on drilling detritus landfill

o Exposure due to cultivation on drilling detritus landfill

Underground construction scenario

Scenario with mine in the vicinity of the repository scenario

7. Evaluation of results and use of probabilities

In Sweden, drinking wells commonly reach a depth of 60 m and so exposure due to utilizing

an intrusion well cannot be ruled out. The drilling scenario calculation with exposure due to

utilizing water from an intrusion well was included in the main risk assessment for which

compliance with the regulatory risk criterion of 10-6

/y (nominally comparable to 14 µSv/y)

needed to be assessed. The doses for the well scenario were relatively high, up to 4.5 mSv.

However, the footprint area of the repository is small and the likelihood of a well in this area


16

is very low. Thus, it was deemed appropriate to assess intrusion wells as a less probable

scenario and assign probabilities. For the majority of the analyzed period, the main scenario

made up the largest risk but around 3000 AD, drilling directly into one of the rock vaults

accounted for the highest risk. The total risk summed over all scenarios was below the risk

criterion of 10-6

/y for the entire assessment period of 100 000 years.

For other drilling scenarios (drilling personnel, construction worker and farmer), the doses

was always low, at most 0.25 mSv. This is well below the ICRP ranges of reference levels

indicative of system robustness (ICRP, 2013). Use of these reference levels is another way of

addressing the generally low likelihood of intrusion without explicit consideration of the

probability. The FHA scenarios, mining in the area and water management work, were

evaluated qualitatively. FHA were considered already in siting and these scenarios were

determined to be unlikely and to have little effect on the repository.

8. Conclusion

An international consensus on how to assess FHA would be very welcome and useful. It is

hoped that SKB work, shared though mechanisms such as HIDRA is a useful contribution to

development of suitable guidance. However, there will always be relevant site, waste type

and repository design factors to take into account when conducting specific safety

assessments, alongside local stakeholder interests and national regulatory requirements.

REFERENCES

[1] NEA, Risks Associated with Human Intrusion at Radioactive Waste Disposal Sites.

Proceedings of an NEA Workshop. Nuclear Energy Agency, Paris (1989).

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal of Radioactive Waste,

Specific Safety Requirements, IAEA Safety Standards Series No. SSR-5, IAEA, Vienna

(2011).

[3] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,

Radiological Protection in Geological Disposal of Long-lived Solid Radioactive Waste,

ICRP Publication 122 (2013).

[4] SEITZ, R., et al., "Role of Human Intrusion in Decision-Making for Radioactive Waste

Disposal - Results of the IAEA HIDRA Project - 16287," Proceedings from the WM2016

Conference, March 6 - 10, 2016, Phoenix, AZ, 201 (2016).

[5] BAILEY L., et al., PAMINA Performance assessment methodologies in application to

guide development of the safety case. European Handbook of state-of-the-art of the safety

assessments of geological repositories – Part 1. Deliverable 1.1.3 (Ch. 9), European

Commission, (2011).

[6] SMITH G.M., et al., Human Intruder Dose Assessment for Deep Geological Disposal.

Report prepared under the BIOPROTA international programme. Available at

www.bioprota.org (2012).

[7] SKB, 2014, Safety analysis for SFR Long term safety. Main report for the safety

assessment SR-PSU. Technical report TR-14-01, Swedish Nuclear Fuel and Waste

Management Co, Stockholm.

http://www.bioprota.org/


17


DATA MANAGEMENT TO SUPPORT A POST-CLOSURE SAFETY CASE FOR

HIGHER ACTIVITY WASTES

A.J. Carter, L.E.F. Bailey

Radioactive Waste Management Ltd., Building 587, Curie Avenue, Harwell Campus, Didcot,

Oxfordshire, OX11 0RH, UK


Abstract. In this paper we describe how RWM has developed its approach to data and model management,

starting from a set of formal ‘aims and principles’; and promoted a culture in which these can operate

effectively. In the course of this development, a number of innovative systems and tools have also been

produced which facilitate the storage and use of data. These will be described, as will lessons learned during the

roll out to date.

Key Words: Data management, Model management, Safety case production.

1. Introduction

The United Kingdom (UK) is committed to the safe management and disposal of higher

activity radioactive waste. This will be carried out through the interim storage of radioactive

waste packages prior to their final disposal in a deep geological disposal facility [1].

Radioactive Waste Management Ltd (RWM) is responsible for the delivery of such a facility,

and maintains a generic Environmental Safety Case (ESC) [2] for UK wastes while a site is

identified through a siting process in partnership with local communities and government.

Following the production of the 2013 UK Radioactive Waste Inventory, the generic ESC is

being updated to take into account changes to waste inventory and packaging, and to reflect

developments in scientific understanding which has resulted from new research since 2010,

when the previous generic ESC was published. The generic ESC is supported by a generic

post-closure safety assessment [3] which presents illustrative calculations to support RWM’s

confidence that a safety case, consistent with the regulatory risk guidance level and other

stakeholder expectations, could be produced in UK-relevant geologies. The probabilistic

computer models underlying these calculations are referred to as total system models (TSMs)

as they contain high-level representations of the total system, that is the wasteform, container,

engineered barrier system, geosphere and biosphere. Significant quantities of input data are

required for these models covering multiple disciplines across RWM’s programme.

In the period since 2010, RWM has undertaken a formal project to review and update its

procedures relating to data and model management. The project has received strong support

from RWM’s Executive team and has resulted in significant improvements in the way data

and models are documented, managed and used across the company. These have now been

applied in the production of the TSMs introduced above and have thereby helped to ensure

the quality and traceability of calculations which are used to support the safety case.

In this paper we describe how RWM has developed its approach to data and model

management, starting from a set of formal ‘aims and principles’; and promoted a culture in

which these can operate effectively. In the course of this development, a number of

innovative systems and tools have also been produced which facilitate the storage and use of

data. These will be described, as will lessons learned during the roll out to date.



18

2. Data Management

An initial step in the data management project involved writing a high-level Policy and

Principles document which recognises the company’s reliance on data and sets out a vision

for the management of its data. This is followed by a set of five principles:

Data and information management is part of everyone’s role;

Data are an asset;

Data and information quality will be assured at source and maintained;

Data and information will be accessible; and

Data and information integrity and security will be assured.

Each principle is used to formally derive a set of implications, and these in turn are used to

inform the underlying data management procedure and define the requirements of its

supporting systems and tools. As an example, the principle that ‘data and information will be

accessible’ implies that each dataset should be made available quickly after its acquisition,

that a means should be provided for staff to discover the dataset, and that a mechanism

should be provided for staff to obtain or access the dataset. The procedure addresses these by

providing flow charts to describe how to register a new dataset, record the use of a dataset

and retire a dataset. The principle also implies a technical requirement to store data using a

sensible file format and attach appropriate metadata to facilitate search. Similarly the

principle that ‘data and information integrity and security will be assured’ implies

requirements for access control, backup, business continuity/recovery arrangements, the

creation of audit trails as datasets are periodically maintained or updated, and the use of

storage locations which minimise the potential for decay or corruption of data.

The accountability for each dataset lies with a senior member of staff from an appropriate

department, known as the data owner. Data owners are accountable to the organisation for the

security, integrity, quality and availability of their data, including making adequate provision

for its long-term care and ensuring it is managed in line with the data procedure. Ownership

of data at a senior level with the organisation helps to reinforce the importance of data

management while ensuring appropriate oversight of data, and its use in business decision

making, at a strategic level. Data owners may delegate their day to day responsibility for a

dataset (for example fielding queries from users) to a data steward, typically a member of

staff within their department, but still retain overall accountability for the dataset.

To support the proper characterisation of a dataset, a number of attributes have been defined:

Characterisation of uncertainty: the degree to which uncertainty, variability and

precision in the data are understood and represented in the dataset;

Provenance: the presence of information within the dataset to describe its source,

underlying assumptions and methods of production; and

Limitations of applicability: the degree to which the dataset addresses the entire scope

of the domain, for example spatial or temporal, together with its internal consistency

(for example in values, terminology, production methodology or in the adoption of

standards).

Each attribute must be populated by a suitably qualified and experienced person who

understands the dataset – ideally at the time of the creation of the dataset – and should be

stored with the dataset. An additional attribute has also been defined to support the proper use

of a dataset:

Relevance: the degree to which the dataset meets the needs of a particular use.

Again this attribute must be populated by suitably qualified and experienced people, typically

recording the agreement between a data owner or steward, who understands each dataset


19

(together with any other datasets which may be relevant), and a data user, who understands

the use the data will be put to (for example a model technical owner, introduced in Section 3,

who understands the underlying conceptual model). In practice agreement from multiple

owners or stewards is likely to be required for a model which covers multiple disciplines.

To achieve its mission to build and operate a geological disposal facility, RWM must

demonstrate that it has the required nuclear safety and environmental competencies for each

stage of process. To support this, a competence management system has been introduced in

order to define and assess whether its staff are suitably qualified and experienced in each

area. The availability of staff with appropriate competencies will also be reviewed as the

siting process progresses and is also used to inform recruitment needs. A competence panel,

chaired by the chief scientific advisor, is used to determine whether a member of staff is

currently regarded as competent in a given area, using an agreed list of requirements for that

area. Evidence of competency, including a list of formal qualifications, skills and experience

are used to make this judgement and are captured via the completion of a competence

assessment form. Any training and development needs which arise from the competence

panel are captured on the development plan for the member of staff, and are reviewed as part

of the performance management process at six monthly intervals. Introduction of a

competence management system has also helped RWM management to identify business

risks, for example skill shortages or key skills which reside with a single member of staff. At

least one of the data owner or data steward must be regarded as competent in the appropriate

technical area for a dataset and this is confirmed as part of the approval process.

Two electronic forms have been developed to support the new data management process,

named a data definition form (DDF) and a data use form (DUF). The DDF is used to define a

dataset, and to record the data quality attributes relating to uncertainty, provenance and

applicability described above. A DUF is used to identify a data need (for example

radionuclide half-lives for the total system model) together with the dataset which will be

used. The decision on the dataset to use takes place through agreement between a data owner

or steward and the data user, as noted above, and is recorded via completion of the relevance

attribute on the DUF, where any caveats or risks on the data of this data for this purpose are

also documented. The DUF for a model which covers multiple disciplines is likely to

reference several DDFs, each potentially with a different data owner or steward, and the

relevance attribute would need to be populated for each DDF used.

The forms are created and edited using a custom .NET based application which saves the

DDF or DUF in an XML compliant format. The application includes the ability to refer to

reference documents stored within the company knowledge base and allows numerical data to

be either directly entered into the form or referenced to an external file (with a secure

checksum used to ensure integrity). Deterministic values and probability distribution

functions (PDFs) are supported and the application is dimensionally aware, so that suppliers

of data may enter each physical quantity in the system of units most appropriate to their field.

An application programming interface (API) has been created so that each DDF (or DUF), or

table of data within a DDF, or individual value may be accessed from code, with filters

available to export the data to GoldSim (used for the total system model) and Microsoft

Excel. This helps to remove transcription errors, and the code is also able to convert each

item of data to a specified system of units thereby removing unit conversion errors. An

additional export filter has recently been produced which is able to inject data into a template

document using the Office Open XML format (for example a .docx file used by Microsoft

Word), thereby allowing automated population of a data report. While the use of code to

populate models and reports carries an initial overhead, subsequent updates are made

considerably easily.


20

3. Model Management

A key component of the updated approach to model management requires the production of a

model register together with a model risk assessment and quality plan (MRAQP) for each

model. The model register, which is easily accessible over the web from the intranet home

page, identifies each model which is used by the company, together with a model senior

responsible owner [4] (MSRO) and model technical owner (MTO), analogous to the data

owner and data steward introduced above. The register also provides hyperlinks to the

MRAQP’s and storage locations for the version-controlled models themselves. ‘Write’ access

to each folder is restricted to the MSRO and MTO to prevent model users changing master

copies, and users are required to consult the register each time they make use of a model to

ensure that they are aware of any updates.

The MRAQP asks the MSRO to describe what the model does, provide a commentary on the

key uncertainties in model results and formally identify the uses of the model (for example by

listing the products results from the model feed into). On the basis of the key uncertainties

and model uses, an overall risk assessment is produced for the model, including an

identification of whether the model is ‘business critical’, and this is used to inform the overall

quality plan for the model. The overall quality plan identifies the level of model verification,

validation and benchmarking required, together with the arrangements for model planning,

version control, design, build and sign-off. It should also give guidance on the extent to

which uncertainties and caveats need to highlighted when presenting model outputs, together

with any other risk mitigations which have been identified. The MRAQP is intended to be a

live document which evolves with the model. It is formally approved by the MSRO and then

made readily available to users of the model or its results.

4. Conclusions

RWM is currently updating its generic ESC to take into account changes to waste inventory

and packaging, and to reflect developments in scientific understanding which have resulted

from new research since 2010, when the previous generic ESC was published. Since this time

the company has reviewed and significantly improved its procedures relating to data and

model management, as well as develop systems and tools to help support this process. Within

this paper, a discussion has been provided to explain how these improvements were

developed, together with a description of the key elements within the system. The updated

total system model which provides illustrative calculations for the new ESC is fully

compliant with the requirements of this system. In future RWM intends to investigate the use

of electronic signatures and electronic workflow to further improve the system.

REFERENCES

[1] DEPARTMENT OF ENERGY & CLIMATE CHANGE, Implementing Geological

Disposal, URN 14D/235, July 2014.

[2] RADIOACTIVE WASTE MANAGEMENT., Generic Environmental Safety Case

Main Report, DSSC/203/01, In publication.

[3] RADIOACTIVE WASTE MANAGEMENT., Generic Post-closure Safety Assessment,

DSSC/321/01, In publication.

[4] HM TREASURY, Review of Quality Assurance of Government Analytical Models:

Final Report, Nick Macpherson, March 2013.


21

03c – 05 / ID 193. Disposal of Intermediate Low Level Waste

GREATER-THAN-CLASS C LOW-LEVEL RADIOACTIVE WASTE

CHARACTERISTICS AND DISPOSAL ASPECTS

H. Arlt, T. Brimfield, C. Grossman

United States Nuclear Regulatory Commission (NRC), Washington, D.C., United States


Abstract. United States regulations (Part 61.55 of Title 10 of the Code of Federal Regulations, Waste

Classification) divides Low-level Radioactive Waste (LLRW) into three classes based on the concentration

levels of certain long-lived and short-lived radionuclides. The three waste classes are Class A, B, and C with

Class C having the higher concentration and/or more long-lived radionuclides than the other two classes.

Greater-Than-Class C (GTCC) waste is LLRW that exceeds the Class C concentration limits and is generally

not acceptable for near-surface disposal. GTCC LLRW corresponds to the low- and intermediate level waste

classes identified in the International Atomic Energy Agency’s Classification of Radioactive Waste General

Safety Guide No. 1. The disposal of GTCC LLRW is associated with greater challenges than other classes of

LLRW due to various waste streams having higher specific activities and higher concentrations of long-lived

radioactivity. The U.S. Department of Energy is responsible for the disposal for GTCC LLRW. The paper

contains insights from a qualitative examination of individual GTCC LLRW streams, disposal methods,

disposal environments, exposure scenarios including by means of inadvertent intrusion and groundwater

transport, and the significant interrelationships between these disposal aspects.

Key Words: Greater-Than-Class C, Low-level Radioactive Waste, waste types, disposal

methods

1. Introduction and Background

United States (U.S.) regulations (Part 61.55 of Title 10 of the Code of Federal Regulations,

Waste Classification, or 10 CFR 61.55) were promulgated to ensure the safe land disposal of

low-level radioactive waste (LLRW). The 10 CFR 61.2 definition of LLRW is based on the

exclusion of other waste streams, i.e., LLRW is defined as “radioactive waste not classified

as high-level radioactive waste, transuranic waste, spent nuclear fuel, or byproduct material

as defined in paragraphs (2), (3), and (4) of the definition of Byproduct material” set forth in

10 CFR 20.1003. The regulations divide LLRW into Class A, B, and C where Class A is the

least radiologically hazardous of the three classes and Class C has the higher concentration

levels of certain long-lived and short-lived radionuclides. LLRW that exceeds the Class C

limit, referred to as Greater-Than-Class C (GTCC) waste, is identified as generally not

acceptable for near-surface disposal although U.S. regulation at 10 CFR 61.55(a)(2)(iv)

allows for disposal in a near-surface facility if approved by the U.S. Nuclear Regulatory

Commission (NRC). The U.S. Department of Energy (DOE) is the responsible U.S. federal

agency for disposing of GTCC LLRW. At this time, there is no disposal capability for GTCC

LLRW; however, the DOE has published their final environmental impact statement [1]

which is an important step in the process towards obtaining GTCC LLWR disposal

capability.

A qualitative examination has provided a more comprehensive understanding of the risks

associated with site characteristics and disposal methods when considering GTCC LLRW

disposal [2]. This paper presents a summary of that examination including aspects that need

to be considered for disposal and also discusses disposal challenges under different

environmental settings and exposure scenarios. Specifically, insights were gained by



22

examining individual GTCC LLRW streams, disposal methods, disposal environments,

exposure scenarios, and the interrelationships between these disposal aspects. Performance

assessments of potential disposal sites containing GTCC LLRW would need to examine these

aspects of disposal. The majority of the information and the data on inventory in this paper

was obtained from Ref. [1] and [3].

2. Intermediate-Level Waste and GTCC LLWR

Intermediate-level waste (ILW) is defined by the International Atomic Energy Agency

(IAEA) [4] as waste that contains long-lived radionuclides in quantities that need a greater

degree of containment and isolation from the biosphere than is provided by near-surface

disposal. ILW contains waste with activity levels above clearance levels as described in Ref.

[5]. ILW may contain alpha-emitting radionuclides that will not decay to a level of activity

concentration acceptable for near-surface disposal during institutional controls. In addition,

ILW does not contain levels of activity concentration high enough to generate significant

quantities of heat by the radioactive decay process and has thermal output that is less than

2 kW·m–3

[6]. ILW is generally recommended for disposal at a depth of between a few tens

to a few hundreds of meters.

The radionuclides, activity concentrations, physical and chemical properties and other

characteristics of GTCC LLRW vary considerably and will influence the appropriate

regulatory approach to the disposal of GTCC LLRW including the depth at which it will be

disposed and a disposal site’s dependence on engineered barriers. However, the majority of

GTCC LLRW is more clearly aligned with IAEA’s definition of ILW than it is with the other

IAEA waste classes due to the properties of long-lived radionuclides in the GTCC LLRW

and that GTCC LLRW generally produces less than 2 kW·m–3

thermal output.

3. Characteristics of GTCC LLRW and Wasteforms

DOE has categorized three GTCC LLRW types: activated metals, sealed sources, and GTCC

Other Waste [1]. GTCC LLRW consisting of activated metals can include irradiated metal

components from reactors such as core shrouds, support plates, and core barrels, as well as

filters and resins from reactor operations and decommissioning [7]. Sealed sources are the

second type of GTCC LLRW and are used at hospitals, medical schools, research facilities,

industries, and universities. A third waste type that is not an activated metal or a sealed

source is referred to as GTCC “Other Waste” based on its differing radionuclides and

concentration levels and can consist of contaminated equipment, rubble, scrap metal, filters,

soil, and solidified sludges [7]. The total stored and projected volume of GTCC LLRW in the

U.S. will be approximately 8,800 m3 (311,000 ft

3) and the projected activity of that waste by

2083 will be 5.92 x 106 TBq (160 MCi) [1].

Activated metals are activated by neutron exposure and have a higher activity level than the

other GTCC LLRW types. Activated metals waste can be subdivided into two categories:

routinely generated activated metal and decommissioning activated metals [7]. The neutron

activation products expected to be most dominant in activated metals at the time of disposal

are C-14, Mn-54, Fe-55, Ni-59, Co-60, Ni-63, Mo-93, and Nb-94. Lower concentrations of

some fission products such as Sr-90, Tc-99, I-129, and Cs-137 and various isotopes of

plutonium are also expected to be present on these materials as surface contamination [8].

The projected total volume of activated metal waste is 2,000 m3 (71,000 ft

3) with 5.9 x 10

6

TBq (160 MCi) of activity, although most of the commercial reactors are not scheduled to

undergo decommissioning for several decades [1].


23

Sealed sources are generally small and the radionuclides are generally enclosed in capsules

made, with very few exceptions, of stainless steel, titanium, platinum or other inert metals

and encompassing several physical forms, including ceramic oxides, salts, or metals. Sealed

sources include Cs-137 irradiators that, unlike the smaller sealed sources, are larger than the

standard 208 liter (55 gallon) drum and would be disposed of individually [1]. GTCC sealed

sources may contain one of many radionuclides including Cs-137, Pu-238, Pu-239, Am-241,

and Cu-244, and their activities can range from 4.07 x 10-4

TBq (0.011Ci) to 1.5 x 105 TBq

(4.1 MCi). The projected volume of GTCC commercial sealed sources is 2,900 m3

(102,000 ft3) [1]. In many cases, the volume includes the device as well as the source since it

may be expeditious to dispose of the device and source as a unit.

The total stored and projected GTCC Other Waste activity is 1.98 x 104 TBq (0.53 MCi) and

relatively small compared to activated metals and sealed sources, although GTCC Other

Waste has a large variety of radionuclides, includes some very long-lived actinide isotopes,

and comprises the largest volume of the three waste types with 3,900 m3 (138,000 ft

3) [1]. A

wide spectrum of radionuclides can be present in this waste type with the isotopes of various

actinides (e.g., uranium, neptunium, plutonium, americium, and curium) being of higher

concern with regard to long-term waste management [8]. GTCC Other Waste generated from

routine operations includes contaminated clothing, floor sweepings, paper and plastic while

decommissioning waste can include building, piping, hardware, and equipment debris.

4. Disposal Aspects

Currently, disposal of LLRW that is not GTCC occurs near the surface with favorable

topographic and geological characteristics and/or with engineered barriers and other features

that impede or limit the eventual release of radionuclides from those facilities. The goal of

disposal is to isolate or limit the release of radioactive waste to the environment for hundreds

to thousands of years. For disposal sites with favorable geological and climatic

characteristics, natural barriers will reduce the number of engineered barriers needed to slow

contaminant release into the groundwater and atmosphere. However, modern disposal

practices can include multi-barrier systems that employ both natural and man-made

engineered barriers. The disposal methods chosen for GTCC LLRW disposal will be critical

to ensuring long-term safety.

In this paper, the lower boundary of near-surface disposal sites is considered to lie 30 m

(circa 100 ft) below the local topographic low point since 30 m is considered the maximum

depth of excavation for the foundations of tall buildings [5]. No generally agreed upon value

to define intermediate depth exists. However, most of the literature uses depths that start at

the near-surface lower boundary and include depths as deep as 100 - 150 m (300 to 500 ft)

under the surface [6] [9]. The disposal methods discussed [1] include disposal in concrete

structures or in trenches near the surface, disposal in borehole and shafts at intermediate

depths, and disposal in a deep geologic repository.

For the deep geologic repository disposal method, disposal sites could be located in semiarid

and arid environments as well as humid environments. However, sites with humid

environments would need to be designed for favorable saturated disposal conditions or be

located in hydrogeological settings that allow relatively dry conditions below water table

elevations (e.g., salt deposits, very compact clay layers, dry bedrock). Intrusion could only

occur if a borehole was drilled very deep. Assuming an inadvertent intruder-driller exposure

scenario was plausible, technical bases and assumptions concerning the degradation of the

stabilizing agent (e.g., grout) and the corrosion rate of metals would be important. For

exposure scenarios involving groundwater transport offsite, performance assessment results


24

[10] indicate that activated metals would contribute more than the other GTCC LLRW types

during any activity concentration release.

For a borehole disposal method at an intermediate depth, disposal sites with humid

environments would be suitable if waste is designed to be disposed in a saturated

environment or if hydrogeological setting allowed waste to be placed in deeper, yet dry,

disposal sites. If an inadvertent intruder-driller exposure scenario is considered plausible,

depth and size of the waste package (e.g., sealed sources) would need to be factored into the

plausibility. Assuming such an exposure scenario is plausible, technical bases and

assumptions concerning the degradation of the stabilizing agent, concrete, and outer canister

of sealed sources would be significant as would the corrosion of activated metals. For GTCC

Other Waste, area concentration limits during disposal may be possible. For the groundwater

transport exposure scenario, the borehole disposal showed the lowest peak dose in

comparison with the trench and vault disposal methods [1].

For a concrete structural containment disposal method and the trench disposal method at the

near surface, disposal sites with humid environments could be suitable; however, additional

engineering controls would be required to address the higher precipitation and potential

erosion rates at these sites. Ref. [1] trench design includes a 5 m (16 ft) minimum cover that

would be deeper than most building construction sites to limit the potential for inadvertent

intruder exposure scenarios. Increased infiltration rates (relative to an arid site) would make

grout, concrete and metal degradation rates especially important. If the GTCC Other Waste

from the West Valley Site in the State of New York was included in the calculations for

groundwater transport at humid sites, GTTC Other Waste would be the main contributor to

peak dose due to the readily soluble nature of this waste type in comparison to activated

metals and seal sources. The radionuclides contributing to peak dose include C-14, I-129,

uranium, and transuranic radionuclides including isotopes of plutonium and americium. For

waste disposal in above-ground concrete structures, the vulnerability to erosional processes

increases potentially allowing more infiltration to occur as the overlying material becomes

less thick and root zones move closer to the barriers. For the semiarid to arid sites, peak doses

were lower [1]. For intruder-driller exposure scenarios, degradation assumptions are again

significant since a driller most likely would not drill through a large intact metallic

wasteform, but would be more likely to drill through a degraded wasteform or degraded

concrete barrier.

5. Summary

A qualitative examination of the challenges associated with GTCC LLRW disposal have

shown how the interrelationships between different disposal site characteristics and the

diverse GTCC LLRW types would make it difficult to regulate the disposal of such waste

within a prescriptive, generic framework: guidelines that may apply to one GTCC LLRW

type may not apply to the other; a disposal method that allows adequate performance in one

environmental setting performs poorly in another. Currently, NRC staff, as the U.S. regulator

for GTCC LLRW, is carrying out a quantitative examination of the different GTCC LLRW

types in context with the many disposal aspects. If NRC staff concludes, as a result of its

analysis, that that some or all GTCC LLRW is potentially suitable for near-surface disposal

with or without special processing, design, or site suitability conditions, NRC staff would

proceed with the development of a proposed rule to include disposal criteria for licensing the

disposal of such waste.


25

REFERENCES

[1] UNITED STATES DEPARTMENT OF ENERGY, Final Environmental Impact

Statement for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive

Waste and GTCC-Like Waste, DOE/EIS-0375, Washington, D.C. (2016).

[2] UNITED STATES NUCLEAR REGULATORY COMMISSION, Historical and

Current Issues Related to Disposal of Greater-Than-Class C Low-Level Radioactive

Waste, Enclosure 2. Technical Considerations Associated with Greater-Than-Class C

Low-Level Radioactive Waste Disposal and Qualitative Examination of Disposal

Challenges, SECY–15–0094, ADAMS Accession No. ML15162A821, Washington,

D.C. (2015).

[3] UNITED STATES DEPARTMENT OF ENERGY, Draft Environmental Impact

Statement for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive

Waste and GTCC-Like Waste, DOE/EIS-0375-D, Washington, D.C. (2011).

[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive

Waste, IAEA Safety Standards Series No. GSG-1, IAEA, Vienna (2009).

[5] ORGANIZATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT \

NUCLEAR ENERGY AGENCY, Shallow Land Disposal of Radioactive Waste:

Reference Levels for the Acceptance of Long lived Radionuclides, A Report by an

NEA Expert Group, OECD, Paris (1987).

[6] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal Approaches for Long

Lived Low and Intermediate Level Radioactive Waste, IAEA Nuclear Energy Series

No. NW-T-1.20, IAEA, Vienna (2009).

[7] BRIMFIELD, T.C., et al., “Not All Greater-Than-Class C (GTCC) Waste Streams are

Created Equal,” Waste Management (WM2015 Conference, March 15-19, 2015,

Phoenix, Arizona, USA), Phoenix, AZ (2015).

[8] ARGONNE NATIONAL LABORATORY, Supplement to Greater-Than-Class C

(GTCC) Low-Level Radioactive Waste and GTCC-Like Waste Inventory Reports,

Washington D.C. (2010).

[9] UNITED STATES CONGRESS, An Evaluation of Options for Managing Greater-Than-

Class-C Low-Level Radioactive Waste,” OTA-BP-O-50, Office of Technology

Assessment, Washington, D.C. (1988).

[10] SANDIA NATIONAL LABORATORIES, Basis Inventory for Greater-Than-Class C

Low-Level Radioactive Waste Environmental Impact Statement Evaluations, Prepared

for U.S. Department of Energy, Washington, D.C. (1988).


26


IMPLEMENTATION OF A GRADED APPROACH IN RADIOACTIVE WASTE

MANAGEMENT IN FRANCE

J.M. Hoorelbeke, S. Thabet

Andra, 1-7, rue Jean Monnet F-92298 Châtenay-Malabry cedex, France


Abstract. Andra is operating near-surface facilities at the industrial scale in France to dispose of very low

level and low level and short-lived wastes. Otherwise Andra’s deep geological Cigéo project is under

preparation to dispose of long-lived ILW and HLW, a large part of them resulting from spent fuel reprocessing.

In between those wastes that can be accommodated by near-surface existing facilities with respect to safety and

those wastes which require the high degree of isolation and containment provided by deep geological disposal, a

wide range of wastes will have to be managed in appropriate disposal facilities to be developed. Some are

legacy while others will be generated in the future. They include for instance radium bearing and other potential

NORM wastes as well as particular decommissioning radioactive waste such as graphite waste (recognized as

“low level long-lived waste” in France). Furthermore the diversity of decommissioning VLLW streams may

suggest dedicated disposal routes in the future.

IAEA’s Specific Safety Requirements SSR-5 provides that the ability of the chosen disposal system for a waste

type to provide its containment and to isolate it from people and the environment is to be commensurate with the

hazard potential of this waste in accordance with a graded approach. Hazards vary widely due to the diversity of

radioactive emission types and energies, of half-lives of radionuclides, of chemical properties and of bio-

toxicity. The needs for isolation and containment are to be formulated in terms of performance, for instance

retardation and mitigation, as well as in terms of a suitable assessment timescale. This timescale is to be defined

consistently with the potential reduction of activity of the waste with time and the evolution of the disposal site

and the containment system.

The definition of an appropriate disposal system includes the natural and/or engineered containment barriers, the

disposal depth, the site characteristics and their evolution over the considered timescale with regard to local

geodynamic conditions, the specific measures that may be implemented during the institutional control period

etc. Social acceptability is a crucial factor in determining proportionate solutions as well as siting disposal

facilities.

Within the framework of the French National Plan for the Management of Radioactive Materials and Waste,

Andra is developing a graded approach to propose new disposal options to complement the existing facilities

and the Cigéo project, in a view to optimizing the use of disposal capacities.

1. Taking into account the diversity of waste for a proper management

Today very low level and low level and short-lived radioactive wastes (VLLW, LLW) are

being disposed of in France by Andra in dedicated near-surface facilities. Otherwise high

level waste (HLW) and long-lived intermediate level waste (ILW) are planned to be disposed

of in deep geological Cigéo project under preparation.

A wide range of wastes may be considered as “in between”: their harmfulness makes them

unsuitable for surface disposal but does not necessary require geological disposal at great

depth. They include graphite waste, waste containing radium and some other waste such as

bituminized sludge from the treatment of effluents in nuclear facilities or maintenance

waste [1]. Most graphite waste comes from the dismantling of former natural uranium gas-

cooled reactors. Radium-bearing waste and broader NORM waste is mostly produced by non-

nuclear industrial activities.



27

An example of the consideration of “in between” wastes is given by the work carried out with

the IAEA from 2013 to 2016 to discuss specifically the disposal of ILW and provide a

reference for selecting appropriate disposal concepts including a suitable depth. Indeed ILW

can be considered as in between LLW that are suitable for near-surface disposal and HLW

that require deep geological disposal. Similar principles may be applied to the broader work

to be carried in France to implement waste management solutions that aim at being

proportioned to the harmfulness of the wastes, consistently with IAEA SSR-5, which

provides that: “In accordance with the graded approach, as required in the International

Basic Safety Standards and other standards, the ability of the chosen disposal system to

provide containment of the waste and to isolate it from people and the environment will be

commensurate with the hazard potential of the waste” [2].

At the lower end of the range of radioactive wastes, most VLLW come from the dismantling.

Their activity level may widely vary and large volumes will arise in the future. A significant

part would be below the clearance level used in a number of other countries but not

considered in France: the disposal of this “very” very low level waste in dedicated facilities

may help to maintain traceability for future generations. Within this framework, the amount

and diversity of VLLW suggest adapting disposal solutions to the specificities of various

VLLW streams.

2. Harmfulness of wastes and needs for isolation and containment

Radioactive waste presents a potential hazard to human health and the environment and it

must be managed so as to ensure any associated risks do not exceed acceptable levels in the

short term as well as in the long term. As pointed out during IAEA technical meetings on the

safe disposal of ILW, hazards vary widely due to varying types of radioactive emissions,

varying energies of these emissions, half-lives of the nuclides in the waste as well as

chemical properties of various contented substances. In addition to the radiological hazard,

waste may also contain chemically toxic components, such as heavy metals. Contaminated

asbestos may also be present in nuclear facilities. Some radionuclides such as uranium

present both a radiological and a chemo-toxic hazard.

Waste management includes a number of successive steps such as sorting, treatment,

recycling as possible, conditioning and storage. End waste is to be disposed of. According to

IAEA safety standards, containment and isolation are the basic principles underpinning safe

disposal of waste to protect man and environment. The choice of a disposal solution needs to

ensure that these principles are met to the degree necessary for the waste during operation

and after closure of the facility. This degree of containment and isolation includes level of

performance as a function of time, taking into account the half-lives, activities and types of

the radionuclides in the waste to be disposed of.

Containment consists in preventing or controlling the release of radioactive substances and

their dispersion in the environment. Isolation is defined in SSR-5 as retaining the waste and

keeping its associated hazard away from the biosphere in a disposal environment that

provides substantial physical separation from the biosphere, making human access to the

waste difficult without special technical capabilities, and restricts the mobility of most of the

long lived radionuclides.

The radiological content of the waste, in terms of half-lives of predominant radionuclides and

in terms of activity level, is crucial to determine the time-scale required for containment and

isolation. Short lived radionuclides are usually considered with a half-life less than around

thirty years. The radiological harmfulness of waste with predominant short-lived

radionuclides significantly decreases within the time scale generally considered for the


28

institutional control (a few hundred years). When considering long-lived radionuclides

defined by half-lives higher than thirty years, a particular attention should be given to the

diversity of these half-lives. Carbon 14, Radium 226 or Americium 241 are predominant in a

wide range of “in between” wastes. Provided the content of these wastes in very long-lived

radionuclides such as Chlorine 36 or Iodine 129 is limited, the required time-frame for

isolation and containment is of the order of 10,000 years. Such a time-frame may be

considered to be long with regard to human civilization, but it is moderate with regard to

geological evolutions, in particular in areas with low geodynamic processes. A much longer

time frame, 100,000 years and more, is required for HLW or spent fuel if considered as

waste:

The required level of performance in containment and isolation of radionuclides is a function

of the type and energy of emission, the activity level in the waste and the mobility of

elements in the geosphere and the biosphere. Regarding isolation in particular, these waste

specific characteristics determine the radiological impact in inadvertent intrusion scenarios as

a function of relating exposure routes (ingestion, external exposure). The needs for isolation

and containment are also to be adapted to the chemo-toxic harmfulness of the waste. Existing

regulation in non-nuclear fields may be used. Radiological and non-radiological harmfulness

should be managed consistently, both in terms of characterization of potential effects on

health and environment and in terms of their consequences on the needs for isolation and

containment. Assessing such a consistency probably requires significant work in the future,

especially when addressing low exposure levels, long time-scales as well as the consistency

with health and environmental protection in non-nuclear activities.

3. Development of proportioned disposal solutions

The first priority before defining disposal solutions is to reduce the volume and the

harmfulness of waste during production process as possible. In France reuse or recycling of

material is also recommended by the French environmental law for any waste – including

non-radioactive and radioactive – as the second priority (“Code de l’environnement”). And

finally end waste is to be properly disposed of. Sorting and treatment can be implemented to

help to reduce volume and/or harmfulness and therefore to facilitate their disposal. Fig. 1

illustrates the main components of waste management as provided by the French National

Plan for the Management of Radioactive Materials and Waste, issued every three years under

the auspices of the Ministry in charge of ecology and the regulator “Autorité de sûreté

nucléaire” [3].

FIG. 1. A typical waste management route according to the French National Plan for the

Management of Radioactive Materials and Waste [3]

Containment and isolation is provided by a combination of natural and engineered

characteristics of the disposal system. This issue has been addressed in detail during IAEA

technical meetings on the safe disposal of ILW, and can be applied to any type of waste:

Containment is achieved by maintaining package integrity, limiting the solubility of

radionuclides and the waste form, minimizing where possible groundwater inflow and/or

providing a long travel time for radionuclide transport from the disposal facility to the

Sorting Treatment Packaging Storage Disposal

Potential transport

Production


29

biosphere; isolation is generally provided by depth of the disposal facility and to some extent

by the geology and environment surrounding the site.

In particular the selected depth of any disposal facility contributes to define the degree and

duration of isolation and of protection from surface erosion due to effects such as glaciation.

Nearer to the surface, natural changes occur over shorter timescales than deeper underground.

Significant processes leading to this evolution include erosion by wind, rain water,

weathering, climate induced processes such as glaciation, etc. These phenomena may change

the future boundary conditions of the system, for example, the hydrographic system and

hydrogeology, as well as the system itself, for example, through the changing chemical,

hydrological and temperature conditions. They will possibly progressively reduce the

thickness and/or performance of containment barriers interposed between the waste and the

environment. In an extreme situation the disposal facility and waste packages may be

destroyed in the long term, leading to loss of containment, direct access to waste and

dispersion of residual activity. The affected depth with time and the speed and consequence

of these mechanisms are site dependent.

The potential contribution to isolation of the institutional control and the memory keeping is

also important, where various time-scales may be considered.

4. Work in progress in France

Andra aims at developing a graded approach to propose new disposal options to complement

the existing facilities and the Cigéo project in connection with suitable predisposal

management options. This approach aims at an overall optimization of the use of disposal

facilities and of the distribution of wastes between these facilities with respect to safety and

cost. It will make it possible to manage all existing and future wastes in a consistent manner,

making the best use of available resources and avoiding undue burden on future generations.

This work requires strong interactions with various stakeholders, including the public. It is

part of the comprehensive approach offered by the French National Plan for the Management

of Radioactive Materials and Waste.

REFERENCES

[1] Inventaire national des matières et déchets radioactifs – www.andra.fr

[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal of Radioactive Waste,

Specific Safety Requirements No. SSR-5, IAEA, Vienna (2011).

[3] French National Plan for the Management of Radioactive Materials and Waste

(PNGMDR) - www.developpement-durable.gouv.fr .

http://www.developpement-durable.gouv.fr/



METHODOLOGY AND RESULTS FOR THE SAFETY ASSESSMENT FOR LOW-

AND INTERMEDIATE LEVEL WASTE REPOSITORY (SFR) IN SWEDEN

K. Källström1, E. Andersson

1, M. Lindgren

2, M. Odén

1, U. Kautsky

1, F. Vahlund

1,

J. Brandefelt1, P. Saetre

1, H. von Schenck

1, P.G. Åstrand

3, P.A. Ekström

3

1 Swedish Nuclear Fuel and Waste Management Company (SKB), Stockholm, Sweden

2Kemakta Konsult AB, Stockholm, Sweden

3Facilia AB, Stockholm, Sweden


Abstract. The Swedish low- and intermediate level waste repository, SFR, has been operating since 1988.

When the nuclear power plants in Sweden will be decommissioned and dismantled additional repository

capacity is required. Additional disposal capacity is also needed for operational waste from nuclear power units

in operation since their operating life-times have been extended compared with what was originally planned.

In December 2014, the Swedish Nuclear Fuel and Waste Management Company (SKB) submitted an

application to the Swedish Radiation Safety Authority (SSM) to extend the existing repository for low- and

intermediate level waste (SFR). SFR, the existing part and planned extension, is placed below the sea floor at

60-120 meter depth in Paleoproterozoic metagranite. For the application an evaluation of post-closure safety is

required. This paper presents the safety assessment performed to evaluate if the repository complies with the

Swedish Radiation Safety Authority’s regulations concerning safety and protection of human health and the

environment in the post-closure perspective. The results from the safety assessment are compared against the

annual risk criterion specified in the regulations, 10-6

, which corresponds to 1 % of the background radiation at

the site. The time frame of the safety assessment is 100,000 years under which there is an evolution of both the

repository and the external conditions (climate and surface systems).

The extended SFR repository and the applied 10 step methodology for the safety assessment are described.

Some steps of the methodology are discussed in more detail, e.g. the FEP-analysis and safety functions.

Different scenarios that will contribute to the overall risk evaluation for the repository are generated from

uncertainties in both external and internal processes. The understanding of the processes is based on extensive

site investigations, research, and numerical modelling of the evolution of the repository and external conditions.

Examples of major results for the dominating radionuclides (C-14, Mo-93 and Ni-59) are presented. The central

conclusion of the safety assessment is that the extended SFR repository meets the regulatory criterion and is

robust and safe in the post-closure perspective.

Key Words: Safety assessment, methodology, SFR

1. Introduction

This paper describes the methodology applied and results from the post-closure safety

assessment preformed to show compliance with the Swedish regulations [1] as a part of the

application to extend the existing Swedish repository for low- and intermediate level waste

(called SFR) situated in Forsmark. The extended SFR repository and the applied 10 step

methodology for the safety assessment are described.



31

2. Description of the repository and the waste

SFR, the existing part and planned extension, is located at 60-120 meter depth in

Paleoproterozoic metagranite below the Baltic Sea floor. Due to land-rise after the last

glaciation, SFR and the overlying rock will in the future be situated below land instead of the

Baltic Sea. The waste is emplaced in different waste vaults consisting of engineered barriers,

adapted to the different protection needs of the particular waste forms. The purpose of the

barriers is to contain the radionuclides, and to prevent or retard the dispersion of those

substances, either directly or indirectly by protecting other barriers in the barrier system.

SFR, with its existing part and the planned extension contains one silo and 10 other waste

vaults (see FIG.1.). The design has been adapted to the properties of the wastes deposited in

each vault. The silo which contains the majority of the activity has both concrete and bentonite

barriers. The two waste vaults 1BMA and 2BMA consist of concrete structures, a waste vault

for boiling water reactor pressure vessels (BRT) is filled with grout, and in two waste vaults

(1BTF and 2BTF) the spacing between containers is filled with grout. For very low level

waste the only barrier is flow limiting plugs installed at closure. For a detailed description of

the repository and the waste vaults see [1].

According to the Swedish regulations the safety assessment for this type of repository needs

to cover a time period of at least 10,000 years but the required time frame is at most 100,000

years. On that time scale the engineered barriers of SFR will, to different degree, degrade.

There are two overall safety principles for SFR – limitation of the activity of long-lived

radionuclides and retention of radionuclides, see Section 2.1.2 in [2]. The activity of the

waste decreases with time, thus relaxing the demands on the protective capacity of the

degrading barriers over time.

3. Safety assessment methodology

The assessment methodology has been further developed since the most recent safety

assessment for SFR, SAR 08 [3], and is largely consistent with the methodology applied in

the safety assessment of the repository for spent fuel, SR-Site [4].The methodology applied

for the post-closure safety assessment SR-PSU consists of 10 main steps illustrated in FIG. 2.

FIG. 1. The existing SFR (light grey) and the extension (blue) with access tunnels. Illustration

reproduced from [2].

1.1 Step 1: Handling of FEPs

This step consists of identifying all factors that need to be considered in order to gain a good

understanding of the evolution and safety of the repository. This is done in a screening of


32

potentially important features, events and processes (FEPs). Experience gained from previous

safety assessments of SFR, including SAR-08 [3], and international databases of relevant

FEPs that affect post-closure safety e.g. the NEA FEP-database [5] are utilised.

1.2 Step 2: Initial state

A thorough description of the waste, the repository and its environs at the time of closure is

needed as a starting point for all further evaluation of the post-closure safety of the

repository.

FIG 2. Overview of the ten steps in the methodology used for the post-closure safety assessment SR-

PSU. Illustration reproduced from [2].

1.3 Step 5: Definition of safety functions

This is a central step and consists of identifying and describing the repository system’s safety

functions and how they can be evaluated with the aid of a set of safety function indicators

that consist of measurable or calculable properties of the wastes, engineered barriers,

geosphere and surface system. The overall safety principles are broken down and described in

terms of a number of specified safety functions and safety function indicators. The fact that a

safety function deviates from its expected status does not necessarily mean that the repository

does not comply with regulatory requirements, but rather that more in-depth analyses are

needed to evaluate safety.

1.4 Step 6: Reference evolution

The purpose of the reference evolution is to provide an understanding of the overall future

evolution of the repository system including the uncertainties of importance for the post-

closure safety of the repository. The reference evolution is an important basis for the

definition of a main scenario and less probable scenarios. The reference evolution covers the

entire time period with an emphasis on the initial 1000 years which is required in Swedish

regulations [6]. The remaining time period consist of temperate climate conditions and

periglacial climate conditions.


33

1.5 Step 8 and 9: Scenarios and calculation of radionuclide transport and dose

With the aid of the safety functions and the description of the reference evolution, a number

of scenarios are chosen to cover possible future evolutions of the repository system. A main

scenario and a number of less probable scenarios are analysed to examine whether the total

risk from all scenarios is below 10−6

. Doses are calculated both deterministically and

probabilistically in coupled box models that include the repository, the rock and the surface

system.

4. Radiological risk

The estimated radiological risks for the main scenario and each of the less probable scenarios

are presented in FIG. 3, The highest radiological risk is generally obtained for the main

scenario, except for a short period around 3000 AD when the highest risk is obtained by

human intrusion in the 1BLA waste vault. Mo-93, C-14, and Ni-59 contribute most to the

total radiological risk but at different time periods. For the entire time period the risk is below

the 10−6 regulatory risk criterion.

FIG. 3. Left: Radiological risk for each scenario taking into account the scenario-specific

probabilities. Right: Contribution to total radiological risk from each radionuclide. Illustrations

reproduced from [2].

5. Conclusions

Due to the combination of sufficiently limited activity of long-lived radionuclides and

sufficient retention of radionuclides in the repository, the central conclusion of the safety

assessment SR-PSU is that the extended SFR repository meets regulatory criteria on post-

closure safety.

REFERENCES

[1] The Swedish Nuclear Fuel and Waste Management Company, Initial state report for the

safety assessment SR-PSU. SKB TR-14-02, Stockholm (2014).

[2] The Swedish Nuclear Fuel and Waste Management Company, Safety analysis for SFR.

Long-term safety. Main report for the safety assessment SR-PSU. SKB TR-14-01,

Stockholm (2014).

[3] The Swedish Nuclear Fuel and Waste Management Company, Safety analysis SFR 1.

Long-term safety. SKB R-08-130, Stockholm (2008).

[4] The Swedish Nuclear Fuel and Waste Management Company, Long-term safety for the

final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project.

SKB TR-11-01, Stockholm (2011).


34

[5] NEA, 2006. Electronic version 2.1 of the NEA FEP database developed on behalf of the

Nuclear Energy Agency by Safety Assessment Management Ltd.

[6] SSMFS 2008:37. The Swedish Radiation Safety Authority’s regulations concerning the

protection of human health and the environment in connection with the final management

of spent nuclear fuel and nuclear waste. Stockholm: Strålsäkerhetsmyndigheten (Swedish

Radiation Safety Authority).


35


IMPLEMENTATION OF REQUIREMENTS ON THE CHEMICAL TOXICITY OF

NUCLEAR WASTE AT A REPOSITORY

A. Glindkamp1, B. Peschel

1, I. Harms

2

1TÜV NORD EnSys GmbH & Co. KG, Hanover, Germany

2Lower Saxony Water Management, Coastal Defence and Nature Conservation Agency

(NLWKN), Hildesheim, Germany


Abstract. In this contribution we will focus on non-radioactive harmful substances in a deep geological

repository. The implementation of specific requirements for the protection of groundwater against pollution is

exemplified by the repository Konrad. We will show how the protection target “protection of water against

pollution” is achieved.

The possible releases of non-radioactive harmful substances via water path were investigated within the scope of

the long-term safety assessment for the repository. Based on this investigation the license for the Konrad

repository was issued including the specific Water Law Permit, which handles the requirements concerning the

possible pollution of groundwater. The specifications of the permit were implemented by the operator of the

repository resulting in an adoption of the waste acceptance requirements.

Key Words: chemical toxicity, repository, groundwater

1. Introduction

In radioactive waste disposal radiological impacts as well as impacts of chemotoxic

components of radioactive waste packages must always be taken into consideration. The

radiological protection target “protection against ionizing radiation” and the protection target

of the near-surface groundwater “protection of water against pollution” have to be

considered.

The repository Konrad is a deep geological repository for radioactive waste with negligible

heat generation (low and intermediate level active waste). The repository is constructed in a

depth of 850 m within an iron ore formation of sedimentary origin, which reveals low, but

existent hydraulic permeability. 400 m of clayey strata above the repository are assumed to

be impermeable and thus form a hydraulic barrier.

2. Long Term Safety Assessment

The safety assessment to evaluate the influences of chemotoxic substances was made by the

Federal Office of Radiation Protection (BfS). It was presented in the IAEA-TECDOC-1325

[1]. The long term safety assessment was based on the scenario that the radioactive waste

with its non-radioactive harmful substances is assumed to come into contact with water

originating from the surrounding rock (‘formation water’) in the post-operational phase and

that non-radioactive harmful substances could be transported into the near-surface

groundwater. To minimize the calculation effort a conservative freshwater model was

applied.

In the meantime the license for the Konrad repository was issued including the specific Water

Law Permit, which is based on the long term safety assessment. The amount of non-

radioactive harmful substance is limited by the Water Law Permit. For 94 substances (e.g.



36

lead, cadmium, toluol) a maximal disposable mass is determined. Further chemotoxic

substances may only be disposed in traces. This means that the quantity of these chemotoxic

substances in the repository is so low, that compromising the near-surface groundwater is

excluded. As it is laid down in the Specific Safety Requirements [2] the associated impact

indicators are given by water specific regulations. In the Water Law Permit it is also

determined that the composition of the deposited radioactive waste has to be monitored.

3. Implementation of the Water Law Permit

To meet these requirements BfS as operator of the repository Konrad has developed a

concept for monitoring the amount of non-radioactive harmful substances contained in the

radioactive waste packages. As one part of the concept, values for the content of non-

radioactive harmful substances in the waste packages (so called declaration threshold values)

are deduced to guarantee that the near-surface groundwater will not be affected.

The declaration threshold values are deduced for each substance, which shall be disposed of

in the repository. In this calculation the relevant limit values in the water specific regulations

are considered as well as substance-specific properties like solubility, composition or

estimated occurrence in the radioactive waste packages. Likewise, it is factored in that

different substances may exhaust the same limit values. For example, iron metal and readily

soluble iron salts both add to the exhaustion of limit value for dissolved iron. Thus, the sum

of the affection of these substances (plus further iron containing substances) on the near-

surface groundwater has to meet this value. The considerably lower solubility of iron metal

compared to readily soluble iron salts leads to a much higher declaration threshold value of

iron metal.

Harmful substances, which are enclosed in a mass fraction below their threshold values, are

classified as trace impurities and can be disposed of without balancing of their amounts. Only

those 94 harmful substances listed in the Water Law Permit can be disposed of in amounts

above their declaration threshold value. A so called material list is generated, in which an

entry for each substance that contains the threshold value and other specifications is

tabulated.

To simplify the description of radioactive waste packages, material vectors can be generated,

which are composed of other entries of the material list. Thereby different waste streams (e.g.

evaporator concentrates, ion-exchange resins) can be described easily by using a material

vector which was generated for this waste stream. In order to describe slightly different waste

streams, variations of the material vectors can be applied for at BfS. For material vectors the

declaration threshold values are deduced on the basis of the threshold values of the contained

substances and their portion in the material vector. Parallel to the material list a container list

is established, in which different containers are described on the basis of their materials.

The responsible water law regulatory authority Lower Saxony Water Management, Coastal

Defence and Nature Conservation Agency (NLWKN) with support of TÜV NORD EnSys

GmbH & Co. KG (TÜV NORD EnSys) as an independent expert organization has evaluated

this concept. It was determined that the concept is suitable to achieve the protection target of

the near-surface groundwater “protection of water against pollution”. Hence the NLWKN

agreed to the concept in 2011.

As a result of the above mentioned Water Law Permit implementation concept, the waste

acceptance criteria for the Konrad repository were adopted. It is now stated that the material

composition of all radioactive waste packages has to be described by the waste owner.

Packages which contain harmful substances above their declaration threshold value can only


37

be disposed of if the contained substance is one of the 94 substances listed in the Water Law

Permit and the amount that is specified there is not exhausted yet.

4. Conclusion

The effect of the non-radioactive harmful substances in the repository Konrad was explored

by the investigation of possible releases via the water pathway in the post-operational phase

of the repository. To ensure that the amount of the non-radioactive harmful substances in the

repository Konrad is low enough to exclude a negative impact on the near-surface

groundwater, so called declaration threshold values were deduced. The evaluation by TÜV

NORD EnSys led to the conclusion that by this approach the protection target “protection of

water against pollution” can be certainly achieved.

By adopting the waste acceptance criteria the waste owners are committed to describe the

material composition of their waste packages. This description can be simplified by using

material vectors and container list entries. Due to the characterized composition of the waste

packages, BfS is able to monitor the materials, which are disposed of in the repository

Konrad, according to the requirements of the Water Law Permit.

The protection of groundwater is an important aspect concerning the disposal of radioactive

waste. The environmental impact of non-radioactive harmful substances should therefore be

investigated taking national regulations for the protection of groundwater into account.

REFERENCES

[1] IAEA-TECDOC-1325, Management of low and intermediate level radioactive wastes

with regard to their chemical toxicity, 2002

[2] IAEA Safety Standards Series No. SSR-5, Disposal of Radioactive Waste, 2011


38


SAFETY ASSESSMENT AS AN INSTRUMENT FOR WASTE ACCEPTANCE

CRITERIA DERIVATION

A. Talitskaya1, E. Nikitin

1, A.Guskov

2, M. Nepeipivo

1, Sh. Garatuev

1, M. Rezchikov

1

1Scientific and Engineering Center for Nuclear and Radioactive Safety (SEC NRS), Moscow,

Russian Federation 2International Atomic Energy Agency (IAEA), Vienna, Austria

E-mail contact of main authors:

[email protected]; [email protected]; [email protected]

Abstract. According to requirements of Russian Federation regulatory framework the substantiation of safety

must be provided in the safety case report. One of the key parts of the safety case is the safety assessment. The

safety assessment must be performed at all stages of a facility lifecycle starting from facility siting and

development of conceptual design until the termination of the regulatory control usually linked to the period of

potential radioactive impact.

The safety assessment performed at designing stage of the near-surface disposal facility for operational and post-

closer period is presented here as a practical example. The main purpose of the safety assessment was a

derivation of maximum total activity and permissible specific activity for considered radionuclides in L/ILW.

Safety assessment for the operational period was performed according to the SADRWMS and GSG-3

methodologies for normal operation, accidental and incidental situations. Performed calculations resulted in the

doses that exceed the safety criteria for staff. Taking this into account permissible specific activity for

considered radionuclides were re-calculated as acceptance criteria.

Safety assessment for the post-closure period was performed according to the ISAM methodology. Normal

evolution scenario and alternative scenarios were considered. Obtained results exceed the admissible level of

radionuclide concentration in ground. Based on proportion of resulted concentration to allowable concentration

in ground the total permissible activity for each radionuclide was re-calculated.

After analysis of both operational and post-closure phases integrated waste acceptance criteria in terms of

radionuclide activity were derived for considered near surface disposal facility.

Key word: safety assessment, safety case, waste acceptance criteria, disposal

1. Introduction

Life cycle of disposal facility goes through several stages, including interrelated operation

and post-closure phases, and according to international practice it is assumed to distinguish

between long-term (post-closure) safety assessment (LSA) and operational safety assessment

(OSA). Operational and long-term safety assessments are widespread and admitted

instruments for objective analysis, assessment of possible radiation impact of radioactive

waste (RAW) disposal facility on human and the environment and decision making.

At the end of 1980th

Back End of the Nuclear Fuel Cycle became one of the most significant

problems of radiation safety for further nuclear energy development. LSA provides

understanding of a facility behavior over a long period. The main purpose of LSA is

estimation and analysis of radiological impact on human and environment due to

radionuclides migration from the RAW disposal taking into consideration wide range of

aspects – geological, chemical, physical, social and others. Our days widely used

methodology was developed within the IAEA Co-ordinated Research Project Improvement of




39

Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities

(ISAM) and then examined and illustrated within the Project on Application of Safety

Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities

(ASAM).Later on it was integrated into Safety Case within the following IAEA Projects:

Practical Illustration and Use of the Safety Case Concept in the Management of Near-

Surface Disposal (PRISM), Practical Illustration and Use of the Safety Case Concept in the

Management of Near-Surface Disposal Application (PRISMA). Result of these projects

became a base for further development of IAEA Safety Standards, such as SSR-5, SSG-23,

SSG-29 and etc. and regulatory documents in the Russian Federation NP-055-14, NP-058-14,

NP-069-14 and etc.

In comparison with long term timeframes of RAW potential hazard, the operational period

and operational safety previously considered as negligible. Only within the International

Intercomparizon and Harmonization Project On Demonstrating the Safety of Geological

Disposal (GEOSAF) it was realized that operational period can significantly affect the long

term safety of disposal facility. At the same time it was recognized in some countries that

safety of disposal facility during operation can’t be demonstrated just by the references to

radiation protection measures and emergency preparedness and response, but should be

somehow numerically assessed and ensured in a systematic manner. In general operation of

disposal facility is close enough to operation of storage facility and it seems to be possible to

use the methodology developed within the IAEA project on Safety Assessment Driving

Radioactive Waste Management Solutions (SADWRMS) and included into the IAEA General

Safety Guide No.3 “The Safety Case and Safety Assessment for the Predisposal Management

of Radioactive Waste” (GSG-3). Similar safety documents are under development in the

Russian Federation.

2. Practical example

For practical purposes one of real Near Surface Facilities for disposal of RAW of classes

3&41 was considered at design stage. The main purpose of the safety assessment was a

derivation of Waste Acceptance Criteria (WAC). Usually only long term (post-closure) safety

is considered for this purpose2 without taking into account operational period of disposal

facility. In this research both operational and long-term safety assessment were taken into

account.

Taken near surface disposal facility is a concrete vault with dimensions (length, width ,

height) - 150 × 25 × 7 m. Annual planned capacity is 1100 m3 of RAW. The whole capacity

of the disposal facility is 22000 m3 according to design. The operational time is supposed to

be at least 20 years. It is planned to place solid conditioned RAW in special concrete NZC

containers. After placing containers in NSF, filling free space by clay powder is assumed to

be performed. The composition of waste radionuclides include: U-238, Cs-137, Sr-90, Co-60.

For preliminary calculations maximum values of specific activity of considered radionuclides

as for RAW of the third class3 according to Russian legislation were used as an input.

1 According to the Governmental Decree No1069…

2 DERIVATION OF ACTIVITY LIMITS FOR THE DISPOSAL OF RADIOACTIVE WASTE IN NEAR

SURFACE DISPOSAL FACILITIES. IAEA, VIENNA, 2003. IAEA-TECDOC-1380 3 10

10 Bk/kg for β-radionuclides, 10

9 Bk/kg for α-radionuclides, 10

8 Bk/kg for transuranic radionuclides


40

2.1.Long-term safety assessment

LSA includes calculation of radiation exposure on the population and the environment caused

by the possible withdrawal of radionuclides from the waste packages and their migration

beyond the safety barriers of disposal facility into the environment after the closure.

Calculations were performed for the maximal period of RAW potential hazard.

As safety indicators the values of specific activities in ground water on the sanitary protection

zone border were chosen. The following assumptions were made: NSF to be constructed,

commissioned, operated and finally isolated in accordance with the design; security,

environmental monitoring and physical control are supposed to be provided during the period

of active institutional control (first 50 - 100 years after the closure); structural integrity of

disposal will be preserved; NSF territory can’t be used by people for living and farming work

during the period of passive control (next 300 years).Normal evolution scenario and

alternative scenarios were considered when performing LSA.

Normal evolution scenario assumes that radionuclides from the waste matrix migrate through

containers, clay backfill and concrete wall of disposal vault into the environment. It was

assumed that the concrete does not change its strength and filtration properties during first

100 years. After 300 years since vault construction, concrete permeability corresponds

approximately to the permeability of sand.

In the period from 100 to 300 years, migration of radionuclides through concrete is due to

convection and diffusion processes, and over 300 years, is determined primarily by

convection. Migration of radionuclides through clay backfill is defined by diffusion process.

After migration through the safety barriers radionuclides get into the unsaturated zone and

further, by filtering with precipitation in the ground aquifer.

The migration of radionuclides in the aquifer is due to convective transport, taking into

account the physico-chemical processes (adsorption, ion exchange, etc.) and molecular

diffusion and hydrodispersion, which will be the scattering factor. As the alternative

scenarios considered "raising the groundwater level". This scenario consider changes in the

hydrogeological conditions at the site through the placement of the disposal 300 years,

despite the fact that the groundwater level rises above the base of the disposal. Because of the

degradation of engineering barriers in the system barriers will be enhanced permeability

zones ("filtration box"). Conservatively assumed that 100% of radionuclides are in the liquid

phase and can migrate with the flow of groundwater to drain, as in the normal evolution

scenario.

On the basis of the developed conceptual and mathematical models calculations using

Ecolego software tool have been conducted.During the LSA uncertainty and sensitivity

analysis were also carried out.

2.2.Operational safety assessment

Main aims of OSA for pre-closure waste management were evaluating of hazards and

radioactive impact on workers, population and the environment.

An individual dose rate for worker equal to 20 m/Sv, and for population – 0,1 m/Sv ware

used as safety criteria. For the environment – air, water and ground concentration (for

accidents and incidents) were used as safety criterion. According to the facility design

following workers are involved into operation of near surface disposal facility during its


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operational period: hoistman, slinger, dosimetrist, controller. The NSF is operated in a shift-

operation mode two times per week. Based on climate statistic it was supposed that 20% of

working days have adverse weather conditions that is why works at these days will be

missed. Total amount of operation modes per year was supposed as 80, average numbers of

containers per one mode is 8. There are 3 configurations of radioactive waste into NZC

container: 100% of Co-60, 10% of Sr-90 + 90% of Cs-137 and 100% of U-238. The

container value is 1,5 m3, wall thickness is 10 cm of concrete. For calculation it was supposed

that at each position works one employee. Next step of OSA was development of normal

operation, incidents and accidents scenarios. NZC protection uptakes α- β- radiation, that was

a reason for Sr-90 and U-238 exclusion from further consideration in normal operation

scenarios. Radiation impact for normal operation is due to external γ radiation of Co-60 and

Cs-137. However, in incident and accident scenarios consideration α-β-radiation may have a

serious impact due to internal exposure. As most dangerous accident scenario was considered

NZC drop with waste release. For each scenarios were developed conceptual and

mathematical models. Based on these models were calculated doses for workers and

population. Dose calculation with consideration direct and scattered radiation.

Operational safety assessment included uncertainties analysis. Uncertainties of time of

procedures may have affection on workers doses during all operational period upto 225%,

uncertainties of workers location relatively to containers – upto 210% and with both

uncertainties – upto 315%.

3. SA results and WAC derivation

Preliminary endpoint results of LSA excess of the safety criteria. Particular, calculations

shows exceeding of specific activity in water on the sanitary protection zone border for

radionuclide U-238 (3.0 Bq/kg according to national requirements for drinking water) when

the initial value of the activity in RAW is 109 Bq/kg. For safe disposal initial specific activity

of U-238 in a container was recalculated for WAC development. After recalculation following

initial activity of radionuclides were obtained: U-238 – 3,0∙105 Bq/kg; Cs-137, Sr-90 and Co-

60 – 1010

Bq/kg (no additional limitation). Preliminary OSA endpoint results also exided

the safety criteria - maximum allowable dose for workers –but for other than in LSA

radionuclides. Dose for public satisfy the safety criteria for normal operation, incident and

accident situations. Specific activity for WAC development were re-calculated based on OSA

results for 3 RAW composition: Co-60 (100%) – 8,94∙107 Bq/kg;

Sr-90 (10%)+Cs-137(90%) – 6,53∙109

Bq/kg; U-238 – 109

Bq/kg (no additional limitation).

OSA and LSA have resulted to different activity restriction. Integrated consideration of Waste

Acceptance Criteria for both LSA and OSA together gives following results:

Co-60 (100%) – 8,94∙107 Bq/kg (based on OSA, no LSA additional limitation);

Sr-90 (10%)+Cs-137(90%) – 6,53∙109

Bq/kg (based on OSA, without LSA additional

limitation); U-238 3,0∙105 Bq/kg (based on LSA, no OSA additional limitation). The

research result shows that just operational either just long-term safety assessment separately

is insufficient for determining those WAC parameters as radionuclide waste composition and

there acceptable specific activities.

4. Conclusion

In general LSA and OSA have similar structure and algorithm. However, scenarios,

instruments, assumptions and models are different. The main impact on WAC from LSA

results is caused by such factors as radionuclides half-life, engineered and natural safety

barriers retardation properties and the migration characteristic of radionuclides. Long-lived


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alpha and beta emitting radionuclides, such as uranium and transuranic elements, C-14 and

Cl-36 have the most impact on safety in long periods. It should be noted that carbon and

chlorine are neutral migrants that is practically not adsorbed by engineering barriers materials

and host rocks.

In case OSA the following factors appeared to be crucial: RAW management system, ,

equipment, number of workers and their qualification, safety culture. Gamma-emitting

radionuclides play the most critical role when considering normal operation. Alpha and beta

emitting radionuclides mainly have no any negative impact during normal operation, while

their presence may have a significant radioactive impact in case incidents and accidents.

According to WAC derivation the results show necessity of both operational and

long-term safety assessment to be carried out on the integrated approach basis. This works

concerns just radionuclide waste composition and there acceptable specific activities WAC

parameters, but there is sharp difference in WAC derivation results with separate

consideration from OSA or LSA standpoint. However, that is just fewer part of parameters and

other parameters derivation needs further researches based on the integrated approach.

Moreover, an integrated approach seems to be essential for other tasks, such as: development

and justification of technical, technological and organizational solutions of disposal;

development and justification of limits and conditions of safe operation and closure of

disposal; development and support of measures aimed at improving the safety of workers, the

population and the environment; justification for changes in the design of disposal etc.


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KONRAD REPOSITORY – EVALUATION OF THE SAFETY REQUIREMENTS

ACCORDING TO THE STATE OF THE ART OF SCIENCE AND TECHNOLOGY

B. Samwer, H. Baumgarten

Federal Office for Radiation Protection

E-mail contact of main authors: [email protected]

1. Introduction – Konrad repository and its history

The Konrad mine, an abandoned iron ore mine located in the area of the city of Salzgitter

(Federal State of Lower Saxony, Germany) is currently being converted to a repository for

radioactive waste with negligible heat generation (Intermediate Level Waste - ILW and Low

Level Waste - LLW). The overall responsibility for the construction and operation of the

Konrad repository is with the Federal Office for Radiation Protection (BfS).

Two shafts were sunk from 1957 to 1962 and the extraction of iron ore started in 1960.

Because of its favourable geology (Figure 1), the mine was investigated for its suitability to

host a repository for LLW and ILW as early as in 1976, after iron ore production had stopped

as a result of non-profitability. The iron ore deposit located in a depth of 1,300 m to 800 m is

12 to 18 m thick. However, the natural barrier in the form of clay and marl layers lying above

the mine is vital; being up to 400 m thick, it seals the mine from groundwater. On account of

the clay and marl layers, Konrad is an exceptionally dry mine, compared with other iron ore

mines.

In 1982, the Konrad mine was proposed as a repository for LLW and ILW with negligible

heat generation. At the beginning of 2007, a definitive plan-approval decision (licence) was

granted for the construction and operation of the repository by the Lower Saxon Ministry for

the Environment (NMU). Thus, the Konrad repository is the first facility for radioactive

waste management in Germany, for which a nuclear plan-approval procedure was conducted

prior to taking it into operation. The Konrad repository is permitted to take up max. 303,000

m³ of radioactive waste with a total activity of β- and γ-emitters of 5.0 · 1018

Bq and α-

emitters of 1.5 · 1017

Bq.

The two shafts of the Konrad mine are about 1.5 km apart. Shaft Konrad 1 serves for

personnel and material transport. Shaft Konrad 2 will serve as emplacement shaft. The

underground situation of the Konrad repository below ground is displayed in Figure 2.

2. Safety analyses for the Konrad mine

Comprehensive safety analyses were made in the scope of the plan-approval procedure for

the Konrad repository. Five aspects of safety analysis were investigated: 1. “Normal

operation”, 2. “Accidents”, 3. “Thermal influence on the host rock”, 4. “Criticality” and, 5.

“Long-term safety”. All safety analyses were examined by experts on behalf of the NMU and

compliance with specifications is controlled also by the state mining authority.



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FIG 1: Geological profile of the region of the Konrad mine showing the iron ore body of a thickness

of 12 m to 18 m. The future repository at a depth of 800 m to 1,300 m is covered by thick clay layers

of up to 400 m.

FIG 2: Underground situation of the Konrad repository. Excavation of the waste galleries has been

completed. Extensive work has been done on the surface facilities of Shaft 1 and the building site

equipment at Shaft 2 was set up.

These safety analyses determine requirements for the technical systems and components, the

operating procedures and the waste packages to be disposed of. They are binding in order to

guarantee safe operation and to minimise possible consequences. Furthermore, it was

investigated in long-term safety analyses how the repository could develop after it has been

sealed and possible consequences were derived. The long-term development of the Konrad

repository was forecast with the help of geo-scientific methods. In model calculations, the

dispersion of radionuclides from the repository up into the groundwater near the surface was

examined and evaluated. The model calculations show that it would take radionuclides at

least 300,000 years to get into the groundwater near the surface. For the transport of long-

lived radionuclides with a higher retention level in the geosphere, the model calculations

show relevant concentrations only after several million years. The calculated maximum

radionuclide concentrations that may occur in the groundwater near the surface have been

taken as a basis for the determination of the radiation exposure in the biosphere. For an

infant, the effective dose calculated according to the provisions set out in the Radiation


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Protection Ordinance is max. 0.26 millisieverts per year (mSv/a); for an adult it is max. 0.06

mSv/a. It is thus lower than the value of 0.3 mSv/a, this value having been applied for

evaluation by the licensing authority. Altogether, the possible impact on the near-surface

groundwater through the release of radionuclides and other pollutants from the repository is

so low that no adverse effects to man and environment need to be feared.

In addition to the safety analysis of normal operation, accidents were analysed. That means,

events in the planned operating procedures which might lead to a release of radioactive

substances into the environment were identified and evaluated. Technical or human failure

and rock-mechanical causes can be the reason for such accidents. In that context, the NMU

stated that the Konrad repository was designed in a manner that is balanced from the safety

point of view. Precaution required according to the state of the art of science and technology

has been taken against damage.

3. Evaluation of the Safety Requirements according to the state of the art of science

and technology

According to the current state of knowledge, there is no information available that is

questioning the safety statements given in the application documents. Furthermore, from the

legal point of view there is no breakpoint for the construction and operation of the Konrad

mine as a repository for radioactive waste. However, the BfS as a responsible owner and

operator has still provided for an evaluation of the safety requirements according to the state

of the art of science and technology prior to the repository being taken into operation. Here

within, the BfS sets a good example to improve safety standards and attempt to contribute to

increase trust and confidence into radioactive waste disposal.

The evaluation of the safety requirements according to the state of the art of science and

technology of the Konrad repository was initiated in 2014 and was continued with an expert

workshop to involve professional audience and stakeholders in April 2016. In the framework

of the workshop, safety-related aspects were collected, discussed and prioritised in three

working groups. The results of the working groups were published on the website of the BfS

and are taken into account in the work of the BfS. Targeted information of professional

audience and stakeholders will be continued at workshops and the public will be informed

continuously about the progress of the work via the internet.

The planned procedure of the BfS includes a step-by-step approach: 1. “Identification of

required updates of the safety analyses” and 2. “Update of safety analyses as required”

(Figure 3). The BfS coordinates and controls the entire process. Preparatory work is ongoing

and the phase, “Identification of required updates of the safety analyses” will be initiated by a

tendering procedure. The BfS will award the contract and the contractor will extensively

assess the safety statements given in the application documents. These safety analyses will be

compared according to the state of the art of science and technology (delta analysis).

Depending on the results, further steps will be conducted. The phase “Update of safety

analyses as required” will be executed if the assessment of the current safety analyses show

deviations from the state of the art of science and technology.

The work of the contractor will be continuously monitored by external experts via scientific

monitoring. The external experts will be installed by the regulatory authority, the Federal

Office for the Regulation of Nuclear Waste Management (BfE). The experts will discuss the

contractor’s results on a regular basis and advise on the work, if needed and the BfS will

coordinate the cooperation of the participants.


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To ensure neutrality and to control quality of the evaluation of the safety requirements, the

contractor’s work will be reviewed comprehensively (Peer Review) at the end of each phase.

The contractor will adapt the work accordingly and the BfS will compile the final results of

the work as well as the outcome of the scientific monitoring and the Peer Review to prepare a

final judgment about the evaluation of the safety requirements according to the state of the art

of science and technology.

In case that evaluation of the safety requirements shows that technical adjustments are

required, the BfS will adjust the planning and adapt possible technical changes prior to taking

the Konrad repository into operation. A periodic evaluation of the safety requirements

according to the state of the art of science and technology will continue after the Konrad

repository has been taken into operation (Figure 3). The entire process will be documented

carefully to prepare guidelines for future safety assessments.

FIG3: Periodic evaluation of the safety requirements according to the state of the art of science and

technology. The planned procedure will be monitored continuously by external experts and the results

will be reviewed comprehensively (Peer Review).

4. Conclusion

The Konrad mine is the first repository in the Federal Republic of Germany which has been

and will be planned, constructed, operated and sealed pursuant to the stringent specifications

of nuclear law, from the beginning of filing the application until the sealing of the mine later

on.

There is no information available that is questioning the safety statements given in the

application documents at this point in time and an evaluation of the safety requirements is not

required by law prior to the repository being taken into operation. However, the BfS as a


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responsible owner and operator proceeds with the assessing of the safety requirements

according to the state of the art of science and technology even now.

Neutrality and transparency need to be ensured throughout the entire procedure and the BfS is

monitored by the federal state of Lower Saxony and by the BMUB.

Further on, additional scientific monitoring of external experts, installed by the regulator BfE

and Peer Review of the evaluation of the safety requirements according to the state of the art

of science and technology is used for the purposes of neutrality. Together with targeted

public relations work, transparency of the process will be promoted to improve public

acceptability of the repository.

Future periodic evaluation of the safety requirements will be ensured by preparing guidelines

based on the current approach. Furthermore, the BfS will continuously monitor the

development of the state of the art of science and technology.

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